ML20203N082
| ML20203N082 | |
| Person / Time | |
|---|---|
| Issue date: | 08/07/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2448, NUDOCS 8609230130 | |
| Download: ML20203N082 (40) | |
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DATE ISSUED: 8/7/86 k "
ACRS SCRAM SYSTEMS RELIABILITY SUBCOMMITTEE MEETING MINUTES JULY 31, 1986 WASHINGTON, DC PURPOSE: The purpose of this meeting is to review the status of the ATWS Rule implementation effort.
ATTENDEES:
Principal meeting attendees included:
ACRS NRC W. Kerr, Chairman W. Hodges J. Ebersole, Member W. Jensen G. Reed, Member (am only)
R. Kendall C. Wylie, Member J. Mauck P. Davis, Consultant BG&E AP&L R. Olson T. Enos W
WOG J. Little M. Burzyski SCE EPRI J. Redmon J. Chao W. Layman An attendees list is attached to the Office Copy of these Minutes.
MEETING HIGHLIGHTS AGREEMENTS AND REQUESTS 1.
Dr. Kerr noted that the ATWS issue has been a major concern of the ACRS over the years. The Subcommittee is interested in learning how implementation is proceeding.
He sees the meeting as infor-mational in nature. He does not anticipate any action by the full Committee at this time. Mr. Reed said that he didn't feel ATWS 8609230130 860807 DESICliATED ORIGIl!AL A
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Scram Systems; Reliability'Minu'tes July 31, 1986 4
_ as a seriou's concern for.some'PWR's.. Mr. Ebersole expressed w
concern for the lack of diversity in the W RPS.
2..
Mr. T. Enos (AP&L) discussed the Program to. improve the performance of the GE AK reactor trip breakers which are common to the B&W and the.CE plants. His presentation focused on the long-term improve-ment program. The' Program's purpose is given in Figure 1.
The Kepner-Tregoe decision analysis method was used to select a course of action for.the breaker reliability problem.
Figures 2-4 detail the alternatives considered and the evaluation criteria used. A study of the reliability problem found fault with:
(1) maintenance practices, and (2) aging of bearing lubricant used on the breakers. The final fixes selected to resolve the above problems are given in Figure S.
The fixes, included:
(1) improved maintenance and surveillance procedures, (2) incorporation of screening and operability criterion, and (3) replacement of trip shaft bearings with Mobil-28 lubricated bearings.
In response to Mr.~Ebersole, Mr. Enos said other devices, besides breakers, were considered to solve the reliability problem. Figure 6 lists the bases for the above fixes. All B&W and CE Utilities have implemented.the above fixes.
Mr. Davis asked if the SCE-proposed fix of adding a second AC shunt trip on the AK-25 breaker was rejected. Mr. Enos said B&W couldn't do this because one set of breakers was DC-powered. Mr. Reed felt the Industry was slowly moving back to use of an energized-to-trip function (shunt trip) as was used on Yankee Rowe.
In response to Dr. Kerr, Mr. Enos said the lubrication problem was masked by the ability of the :. hunt trip to override the seized bearing where the undervoltage trip alone could not trip a breaker in this condition.
Mr. Wylie said there have been problems with binding of breakers
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Scram Systems Reliability Minutes July 31, 1986-after having been bench tested successfully. Mr. Enos said, in re-sponse to Dr. Kerr, that he would check out this issue, as he was not aware of this problem.
Mr. Enos showed breaker test data to illustrate that the above fixes have been successful. During questions, it was noted that the shunt trip device is considered more reliable and, if not required by NRC regulation, the UVTAs would be removed or replaced with shunt trips.
Mr. Enos reviewed the activities of the B&WOG in implementation of the ATWS Rule requirements.
Figure 7 lists the Rule's require-ments.
The 13&WOG ATWS analysis showed that the total loss of feedwater is the most serious ATWS event. The peak pressure for this event, assuming DSS fails and AMSAC works, is ~3464 psig.
In response to Mr. Ebersole, Mr. Enos said the RPS is analyzed for 4000 psig stress, but it is assumed that one would be left with a SB LOCA situation in the above case.
In response to Dr. Kerr, Mr. Enos characterized the analysis as " realistic".
Mr. Enos reviewed the analysis for the selection of trip setpoints for the DSS and AMSAC systems. Given the design bases imposed on the above systems (Figures 8-9), the OG concluded that:
(1)high RCS pressure or loss of main feedwater is an acceptable signal for DSS, and (2) loss of main feedwater is an acceptable signal for AMSAC.
The B&WOG has developed a functional requirements document for the affected utilities' Rule implementation.
Figure 10 shows the current implementation schedule for the B&W plant utilities. The
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Scram Systems Reliability Minutes July 31, 1986-4 emergency operating procedures for ATWS (Figure 11) are in place at I
B&W plants.-
- Dr. Kerr asked if B&W Owners have evaluated the IDCORE ATWS core melt contributor evaluation for applicability to their respective 6
plants. Mr. Enos said the ATWS contributor is considered small for B&W plants. Dr. Kerr asked what value of.unreliability wculd the B&WOG consider acceptable for the RPS. Mr. Enos did not have an
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Dr. Kerr said he does not believe one can verify the RPS answer.
unreliability value specified by NRC for the ATWS problem (
10-5/RY). He also said the downside risk of the ATWS-Rule modi-fications must be evaluated.. He urged the Owners to carefully evaluate.the issues associated with downside risk vis-a-vis the Rule requirements.
-3.
R. 0lson.(BG&E) discussed the status of ATWS Rule implementation for CE reactor plants. Currently, the CE0G is in the process of obtaining clarification from NRR on the acceptability of the OG's 4
proposed ATWS modifications.
Forthe" older"CEplants(pre-St.
Lucie, Unit 1) the modification proposals are considered satis-i factory.
For newer plants (St. Lucie, Unit 2 and on) NRR has accepted the DSS and turbine trip modifications, but the diversity of the existing AFW actuation system is still under Staff review.
i In response to Mr. Ebersole, Mr. Olson said that if the DSS success-fully operates no CE plant will exceed 2600 psig.
If DSS fails, higher ( > 4200 psig) pressure could result in stretching of the head bolts in some CE plants.
Mr. Kapinos (CE) detailed the designs of the hardware for Rule compliance.
Figures 12-15 outlines the DSS and turbine trip fixes.
In response to Dr. Kerr, the CE OG indicated PRA studies show CE AFW system reliabilities in the "high" range (10 10-5 reliabil-l ity/ demand). The CEOG also noted that ATWS emergency operating I
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Scram Systems Reliability Minutes July 31, 1986 guidelines have been developed for use by the plants for incorpo-ration into procedures (Figure 16).
Dr. Kerr asked for the CE plant implementation schedule. Mr. Olson referred him to the NRC.
4.
Mr. R. Kendall-(NRC-NRR) discussed the status of NRR's review of the B&W and CE ATWS implementation effort. He noted that the QA aspects of the implementation effort will be reviewed by the Regional offices. The hardware aspects of the effort will be reviewed at Headquarters.
In response'to Dr. Kerr's questions, Mr. Olson said his understanding was that QA and control aspects of the actual hardware installation would be reviewed on the Region level. The Regional review would focus on the* documentation involved.
Regarding the ATWS E0P's, Mr. Kendall said NRC has not yet complet -
ed its review of these procedures. NRRs review of the RTB mainte-nance and surveillance procedures is also still on going. Dr. Kerr asked if NRC has questioned its putting heavy emphasis on mainte-nance and surveillance (M&S) on the UVTA vis-a-vis the shunt trip.
Mr. Kendall said the shunt trip is included in the M&S for the breaker. Mr. Olson (BG&E) confirmed this. Mr. Olson also noted that the ATWS precedures in use have been reviewed by both the licensees and the NRC.
NRR review of B&W and CE plant specific hardware fixes are scheduled to be completed by the end of FY 1988.
In response to Dr. Kerr, Mr. Kendall said he would provide the currently available dates for completion of the ATWS implementation effort for the CE plants. As a result of further discussion, Dr. Kerr indicated that in his view diversity should be of secondary concern to reliabil-ity.
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Scram Systems Reliability Minutes 6-July 31, 1986 In response to Mr. Davis' question, T. Enos said there is no conflicting requirements for the AFW initiation vis-a-vis SG overfill concern and AMSAC AFW initiation.
5.
J. Redmon (SCE) detailed a proposal to upgrade the reliability of the AK-25 RTB's by addition of an AC shunt trip in lieu of the UVTA.
He prefaced his remarks by stating that he is speaking on behalf of himself.
As a result of the Salem event and repeated problems seen with the AK-25's at the SONGS-2 and 3 units, SCE investigated a design
- change to improve RTB reliability. They settled on change-out of the UVTA with an AC shunt trip to complement the DC shunt-trip already installed in the breaker. Test data shown by Mr. Redmon indicate that the shunt trip has sufficient force to overcome the increased trip torque required even if all three bearings in the breaker are frozen.
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Use of an AC shunt in lieu of a UVTA still retains a fail safe function since loss of AC power will result in rod fall upon loss of the MG set power to the control rod drive mechanisms.
Figure 17 shows the equipment modifications required to install the AC shunt.
Figure 18 details the desirable design features. The undesirable features are that continuity must be assured for the AC supply circuit and the shunt, thus continuity detection may be required.
The NRC Staff is divided on the acceptability of the above r
modification from a licensing standpoint.
(Note:
Since the initial SCE proposal, the CEOG decided on the RTB program noted above and no further action was taken on this item.)
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Scram' Systems Reliability Minutes July 31, 1986-6.
M. Burzynski representing the W0G discussed their effort to-imple-ment-the ATWS Rule requirements. He began by detailing the three
' generic AMSAC logic' designs for use by the W' plants as they choose..
(Note: AMSAC'was the only hardware fix' required by the ATWS Rule for W plants.) Figures.19-21 show the layouts for these three
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designs. NRC has reviewed and approved these designs; SERs have been issued.
t The WOG sees no need to. change the emergency response guidelines j
(ERG) because of the ATWS Rule requirements. NRC approved the WOG ERG's in July 1986.
In response-to Dr. Kerr, Mr. Burzynski said the UVTAs are replaced after 1250 duty cycles. Experience from the RTB surveillance program is showing that testing a breaker every 2 months appears to be an optimum surveillance schedule and that no I
breaker wearout is being seen as a result of the above test schedule.
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'The WOG noted that, based on sensitivity studies performed, if one j
PORY was blocked, peak pressure would remain below ASME stress-
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Level C for the loss of normal feedwater.and loss of load ATWS.
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was also noted that with a nominal MTC value,. Service Level C limits are not exceeded with two blocked PORVs.
In response to Mr.
Ebersole, W said that Service Level C pressures do not result in disabling equipment required for recovery from an ATWS.
RTB experience with an automatic shunt trip attachment has shown j
that the failure rate is reduced by a factor of two. However, there is insufficient data to draw any solid conclusions concerning the improvement in reliability.
In response to Dr. Kerr, Mr.
f Little clarified a prior statement that there has never been a l
failure of a shunt trip to open (emphasis added) a breaker on demand. W also noted that the RTB unreliability rate is in the range of 10-4 failures / demand.
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Scram Systems Reliability Minutes July 31, 1986 Mr. J. Little (W) provided W comments on the EPRI NSAC 91 Report.
Key points made by W_ include:
- For W plants the most limiting criterion is the ASME Service Level C stress limits (3200 psig).
- NRC has reviewed and concurred that W codes are acceptable for evaluation of ATWS transients.
- W analysis case for ATWS assumed BE design conditions and showed that for worst-case ATWS events (loss of load, loss of feedwater) peak pressures were -7 Service Level C limits (3200 psig). The NSAC-91 results for a similar case showed a 3101 psig peak pressure.
Dr. Kerr said the results indicate to him there are significant differences in the two models. He asked which (if either) model he should believe. Mr. Little said the W model compares well with NRC and Laboratory analyses. W also said there is a significant difference in the steam generator heat transfer models used, thus the methodologies are different in approach.
It is these differences in the approach and methodologies used that explain the differences in results seen. Dr. Kerr expressed skepticism with the above conclusion.
In response to Mr. Davis, Mr. Little said that going to longer lived cores would result in higher peak pressures for an ATWS due to the more positive MTCs that would result.
7.
W. Layman (EPRI) provided comments on their NSAC/78 and /91 Re-ports. Mr. Layman reviewed the EPRI generic safety analysis program that resulted in the issuance of the above NSAC Reports.
Figure 22 shows the steps involved in the Program. He said the objective of the /91 Report was to provide independent BE analysis
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~ Scram Systems Reliability Minutes July 31, 1986 of a problem (ATWS) that could be reviewed by the' Industry at large.
J. Chao (EPRI) reviewed the technical points of the NSAC/91 and /78 Reports.
Key points noted included:
In response to Dr. Kerr, Dr. Chao said the code needed to be adjusted as a result of the above validation runs. This was necessary due to a lack of sufficient LOFT test data (Figure 23 - arrows).
- Because the MTC value was unavailable, EPRI fitted their power curve to the data.
- EPRI concluded that:
(1) variation in SG modeling causes 100-200 psi uncertainty in peak pressure; (2) peak pressure is sensitive to initial MTC value; and (3) optimum turbine trip time is 30 seconds.
- Dr. Kerr asked why EPRI believes their results differ from the W analysis. Dr. Chao echoed the reasons cited by Mr. Little.
Dr. Kerr asked which code he should believe. EPRI indicated that the codes' are basically within 200 psi of each other.
Further discussion resulted in no real answer to the question (i.e.,"takeyourpick").
8.
NRR comment on the differences between the W and the NSAC/91 report analyses were discussed by W. Jensen.
Key points noted by Mr.
Jensen were:
- The Staff has not to altered its views on the adequacy of the Westinghouse ATWS model.
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- NSAC-78 indicates abrup't loss of steam generator heat transfer for EPRI RETRAN relative to LOFT data.
- The Westinghouse model was reviewed and audited as described in NUREG-0460.
- Staff audits of CE and Westinghouse codes indicate consistency in methodology.
UK RETRAN analytical results are consistent with Westinghouse.
Mr. Davis asked if NRR is concerned with the trend to lower MTCs associated with higher burnup cores. No ready answer was given other than to acknowledge the trend.
9.
J. Mauck (NRR) discussed the NRR review of the W ATWS Rule imple-mentation effort. Currently, NRR anticipates completion of ATWS Rule requirements by July 1989. NRR issued their SER approving the AMSAC generic designs on July 7, 1986.
Figure 24 shows the current status of the plant specific responses to the Rule requirements.
In response to Dr. Kerr, L. Crocker said he would let P. Boehnert know if plant specific approval will be needed from NRR.
Mr. Hodges noted that GE has submitted a generic topical report which has been subscribed to by 50% of the BWR plants.
NRR's goal is to minimize the plant-specific review to the greatest extent possible.
Figure 25 shows the ATWS Rule requirements for BWRs. The generic options for satisfying the Rule are listed in Figure 26.
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E Scram Systems Reliability Minutes July 31, 1986 NRR reviewed the current status of implementation of the Rule requirements.
Figures 27-29 provide the specific details.
- 11. Dr.-Kerr said he will make an oral. report to the Committee at the August meeting.
He solicited any written comments from the Subcom-mittee Members and Consultants they care to make.
- 12. The meeting was adjourned at 3:57 p.m.
5 NOTE:
Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capital Street, Washington, DC 20001, (202) 347-3700.
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LONG TERM IMPROVEMENT PROGRAM EVALUATION PROGRAM PURPOSE:
e IDENTIFY ROOT CONCERNS WITH GE AK2 REACTOR TRIP BREAKERS e
IDENTIFY ALTERNATIVES FOR CORRECTIVE ACTION i
e EVALUATE ALTERNATIVES e
RECOMMENDED BEST CORRECTIVE ACTION (S)
ALTERNATIVES EVALUATED REPLACEMENT e
W DS206 BREAKER e
W DS416 BREAKER e
LOAD CONTACTORS e
SOLID STATE BREAKERS e
OTHER REPLACEMENT BREAKERS MODIFICATION
- e SECOND SHUNT TRIP DEVICES e
B0OSTED (HOTSHOT) UV/LV DEVICES WITH MANUAL & AUTO RESET e
REPLACEMENT OF TRIP SHAFT BEARINGS WITH MOBIL 28 LUBRICATED BEARINGS
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EVALUATION CRITERIA l
REQUIREMENTS i
e TRIP FROM RPS/PPS e
ADEQUATE ELECTRICAL RATING e
ADEQUATE INTERRUPT CAPABILITY
- e UNRELIABILITY < 10-3/ DEMAND e
RESPONSE TIME LESS THAN SAFETY ANALYSIS REQUIREMENTS I
e COMPLIANCE WITH GDCs i
e QUALIFIED e
DIVERSE FROM SCRs*
e TRIP FORCE GREATER THAN EXISTING **
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SPARE PARTS AVAILABLE TO 2006**
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- B&W UNITS ONLY
- DESIRABLES IN BWOG EVALUATION
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i EVALUATION CRITERIA DESIRABLES TO BE MAXIMIZED e
MINIMIZE RESPONSE TIME e
2/1 TRIP FORCE RATIO e
PARTS AVAILABLE TO 2006*
e MINIMIZE MAINTENANCE COMPLEXITY e
MAXIMIZE MAINTENANCE INTERVAL e
MINIMIZE COST e
MINIMIZE IMPLEMENTATION SCHEDULE e
LICENSABILITY e
MAXIMIZE SIMPLICITY OF DESIGN i
e MAXIMIZE EXPECTED LIFE **
e CAPABILITY TO PROVIDE TRIP CONFIRM **
e MINIMIZE FAILURE MODES ***
- REQUIREMENT BY CEOG
- BWOG ONLY
- CE0G ONLY t
j RESULTS OF FINAL EVALUATIONS j
FINAL MODIFICATION SELECTION j
e J/ IMPROVED MAINTENANCE & SURVEILLANCE PROCEDURES e 3 INCORPORATION OF SCREENING & OPERABILITY CRITERION
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e FOLLOWUP EFFORT TO VERIFY FIX
4 BASIS FOR FINAL FIX e
MEETS ALL EVALUATION REQUIREMENTS i
e RESULTS OF MAINTENANCE IMPROVEMENTS e
UNRELIABILITY LESS THAN 10-3/ DEMAND o
SYSTEM UNRELIABILITY (FAIL TO TRIP) e HIGH RELIABILITY OF SHUNT TRIP e
LUBRICATION TESTING e
COST EFFECTIVE SOLUTION e
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DSS (DIVERSE SCRAM SYSTEM) e NON-SAFETY RELATED e
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' ACCEPTANCE CRITERIA PEAK RCS PRESSURE LIMITED TO 3250* PSI o PRESSURE LIMIT OF ANALYSIS ' CODE USED COMPUTER CODE B&W'S TRAP 2 DIGITAL COMPUTER CODE e BETTER MODELING CAPABILITY OF 177-FA PLANTS FOR PARAMETERS OF INTEREST F
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- ANALYZED LIMIT = 4000 PSI ARBITRARILY CHOSEN FOR SETPOINT EVALUATIONS.
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1 CURRENT IMPLEMENTATION SCHEDULE FOR THE B&WOG MEMBER UTILITIES AP&L (ANO-1) 1988*
DPC0 (0-1, 0-2, 0-3) 1989 FPC (CR-3) 1989 GPUN (THI-1) 1988*
SMUD (RANCHO SECO) 1990 TED (D-B) 1990
- CONTINGENT UPON NRC SER ISSUED BY AuousT 1986.
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EMERGENCY OPERATING PROCEDURES INVOLVING ATWS e
B&WOG HAS DEVELOPED SYMPTOM-BASED EMERGENCY 4
OPERATING PROCEDURES "ATOG" l
e UNDER THE ATOG PHILOSOPHY, REACTIVITY CONTROL IS THE FIRST OF 5 BASIC CONTROL FUNCTIONS WHICH ARE USED TO AFFECT 3 BASIC SYMPTOMS OF HEAT-TRANSFER UPSET.
l VERIFICATION OF REACTIVITY CONTROL SYSTEM STATUS IS THE FIRST OPERATOR ACTION FOR THE ENTRY CONDITIONS UNDER ATOG AND RECOGNITION AND TREATMENT OF ATWS IS THE FIRST "SPECIAL CONSIDERATION" FOR TRANSIENT MITIGATION.
e E0PS ARE USED WHENEVER CONDITIONS EXIST FOR REACTOR TRIP - INCLUDING THOSE CASES WHERE REACTOR TRIP HAS NOT OCCURRED.
UNDER ATOG THE OPERATOR WILL:
DETERMINE THAT REACTOR POWER IS DECREASING ON THE INTERMEDIATE RANGE NI.
IF NOT, MANUALLY TRIP REACTOR.
IF UNSUCCESSFUL, THE OPERATOR IS DIRECTED TO MANUALLY OPEN REACTOR TRIP BREAKERS OR OTHER UPSTREAM BREAKERS.
l SUPPLEMENTARY MEASURES DIRECT THE OPERATOR TO:
ATTEMPT TO DRIVE THE CONTROL RODS INTO THE REACTOR AND INITIATE EMERGENCY BORATION.
e LONG-TERM ATWS RECOVERY IS TREATED LIKE ALL OTHER HEAT TRANSFER UPSETS e
PLANT-SPECIFIC IMPLEMENTATION:
ALL OPERATING PLANTS HAVE IMPLEMENTED PROCEDURES BASED ON THESE GUIDELINES.
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GUIDELINES p.g. 3._ of a vision 1
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Verify reactivity control 2.
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following as necessary:
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- a. Manually trip the' reactor l
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- 6. Open the reactor trip breakers and
- c. No more than one CEA
- c. Deenergize the CEA motor botton light not lit.
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- d. [ Plant specific methods, inserthere)
- e. E more than one CEA not inserted, Then borate the i
plant in accordance with technical specifications.
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ADD LIGHT IN SERIES WITH SHUNT TRIP FOR CIRCUIT MONITORING FM. Ik
r DISCUSSION OF DESIGN DESIRABIE FEATURES:
LARGE TRIP FORCE LONG RELIABLE PERFORMANCE HISTORY WILL TRIP RTB's ON LOSS OF 1E DC CONTROL POWER INFREQUENT MAINTENANCE G0/NO GO SURVEILLANCE GUARANTEE OF SUFFICIENT FORCE TO OPEN RTB's AND NOT REVISITING PROBLEM DOES NOT MODIFY BREAKER STANDARD PRODUCTION DEVICE GENERIC FIX d-
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(ifNecessary)
Inlttale Aummary Feedwater Flow I
l Trip Twbine 4
l l
LOGIC DIAGRAM-FEEDWATER PUMP AND VALVE STATUS s
s r
9)f F
MAINTAIN UP-TO-DATE TECH-NICAL STATUS AND CONTACT WITH INDIVIDUALS AND
--TRACKING ORGANIZATIONS WORKING ON ABOUT 30 GENERIC ISSUES.
i TECHNICAL REVIEW OF GENERIC PROBLEMS.
EVALUATE WORK CURRENTLY BEING DONE TO ACHIEVE RESOLUTION.
--EVALUATION IDENTIFY ADDITIONAL WORK GENERIC REQUIRED.
PREPARE POSITION SAFETY PAPER FOR EACH HIGH PRIORITY ANALYSIS ISSUE.
ANALYSES AIMED AT RESOLUTION OF THE ISSUES WHERE POSSIBLE,
--SAFETY ANALYSIS ARE DONE JOINTLY WITH PARTICIPATION FROM AFFECTED UTILITIES.
- - ~,,., - _ _. _ _, _ _ _,,, _ _ _ _ _ _
LOFT TEST L9-3 V
,1 m. <e..,..
~
/
=
/
Si f,tt.Ag (C *6.%)
e
_ f M
h
~
=tPI,lMMY g
=
d 20 40 60 80 100 IN 140 160 TIE (SEC)
Steam Generator Pressure for LOFT L9-3 1
I I
~
/#a ::th
REVIEW OF PLANT SPECIFIC RESPONSE TO THE ATWS RULE STATUS AS OF 7-15-86 coeec***************************************************************************;l NO.!
WESTINGHOUSE lFINjDOCKET l TRIP l EICSB/DPA ACTION l # jNUMBER lSYSTl--------------------------------------ll l
PLANTS l
l l
l TAC l SUBMIT l
___l____________________l___;_______;____;________l_________;_________;l 1 l BEAVER VALLEY 1 l17 1334 l
l59070 l
l l
2 l BEAVER VALLEY 2 118 l412 l
l l
l
.l l
3 lBRAIDWOOD 1,2 (04/5)l18 l456/4571 l
l l
l*
l 4 l BYRON 1,2 (04/5)l17 l454/455j l59077 l
l l
l 5 lCALLAWAY l17 l483 l
l59078 l
l l
l 6 l CATAWBA 1 l17 l413 l
l59081 l
l l
l 7 l CATAWBA 2 l18 l414 l
l l
l l
l 8 jCOMANCHE PEAK 1,2 l18 l445/446l l
l l
l l
9 lC00K 1,2 l17 l315/316l l59082/3 l l
l l
10 l0IABLO CANYON 1,2 l17 l275/323l l59088/9 l l
l l
11 lFARLEY 1,2 l17 l348/364l l59092/3 l l
l l
12 lGINNA l17 l244 l
159098 l
l l
l 13 l HARRIS 1 l18 1400 lFP/V; l10-14-85 l l
l 14 l INDIAN POINT 3 l17 l286 l
l59104 l
l l
l 15 lKEWAUNEE l17 l305 l
159105 l
l l
l 16 lMCGUIRE 1,2 l17 l369/370l l59111/2 l l
l l
17 l MILLSTONE 3 l18 l423 l
l l
l l
l 18 l NORTH ANNA 1,2 117 l338/339l l59117/8 l 4-10-86 l l
l 19 l POINT BEACH 1,2 l17 l266/301l l59128/9 l l
l l
20 l PRAIRIE ISLAND 1,2 l17 l286/302l l59130/1 l l
l l
21 l ROBINSON 2 l17 l261 lFP/Vj59135 l10-14-85 l l
l 22 l SALEM 1,2 l17 l272/311l l59136/7 l l
l l
23 l SAN ONOFRE 1 l17 l206 l
159138 l
l l
l 24 lSEABROOK 1,2 118 l361/362l l
l l
l l
25 lSEQUOYAH 1,2 l17 l327/328; l59141/2 l l
l l
26 l SOUTH TEXAS 1,2 l18 l498/499l l
l 5-30-86 l l
l 27 l SUMMER l17 l395 l
l59146 l
l l
l l
28 lSURRY 1,2 l17 l280/281l l59147/8 l 4-10-86 l l
l 29 l TROJAN l17 l344 l
l59152 l 3-17-86 l l
l l
30 ! TURKEY POINT 3,4 l17 l250/251l l59153/4 l l
l l
31 lV0GTLE 1,2 l18 l424/425l l
l l
l l
l 32 l WATTS BAR 1,2 l18 l390/391l l
l l
l l
33 l WOLF CREEK l17 l482 l
l59157 l
l l
l 34 l YANKEE R0WE l17 l029 l
l59159 l10-15-85 lWANTS AN l EXEMPTION l 35 l ZION 1,2 (SERIES 51)l17 1295/304lLSGLl59160/1 l 6-5-86 l
l l
TRIP SYSTEM
- 1. LOW-LOW SG LEVEL LSGL j
l
- I l
$* 5f
~
0 10 CFR 50,62 REQUIREMENTS FOR BWRs AUTOMATIC RECIRCULATION PUMP TRIP DIVERSE REACTOR TRIP SYSTEM STANDBY LIQUID CONTROL SYSTEM PROVIDES EQUIVALENT OF 86 GPM a 13% (WEIGHT) 0F SODIUM PENTABORATE AB0VE SYSTEMS MUST MEET THESE REQUIREMENTS:
POWER SOURCE INDEPENDENT FROM CURRENT RPS MUST FUNCTION W/0 0FFSITE POWER MUST NOT DEGRADE ANY INTERFACING SAFETY GRADE EQUIPMENT TESTABLE AT POWER QUALIFIED FOR TRANSIENT CONDITIONS (A00)
NOTE:
NOT REQUIRED TO BE SAFETY GRADE OR REDUNDANT 4
A
o O
OPTIONS PROPOSED BY BWR OWNERS GROUP (TOPICAL REPORT NEDE-31096-P)
RECIRCULATION PUMP TRIP "0RIGINAL" BWR/4 DESIGN MODIFIED HATCH DESIGN MONTICELLO DESIGN DIVERSE REACTOR TRIP SYSTEM ALTERNATE R0D INJECTION (ARI)
STANDBY LIQUID CONTROL SYSTEM TWO PUMP OPERATION IS0 TOPIC ENRICHMENT (B10)
INCREASED S0DIUM PENTABORATE CONCENTRATION l
IfH.5l
O-STATUS OF IMPLEMENTATION RECIRCULATION PUMP TRIP
, APPROVED DESIGN INSTALLED AT 27 UNITS i
THREE UNITS COMMITTED TO UPGRADE 7 UNITS WITH "0RIGINAL" DESIGN NEED UPGRADING BIG ROCK GRANTED EXEMPTION i
.ARI INSTALLED AT 14 PLANTS i
10 OTHERS SCHEDULED BY END OF 1987 8 OTHERS SCHEDULED IN 1988 APPLICABILITY OF RULE QUESTIONABLE FOR LACROSSE BIG ROCK T0' REQUEST EXEMPTION BROWNS FERRY, SHOREHAM SCHEDULES UNKNOWN BECAUSE'0F STARTUP DATE UNCERTAINTIES l
O STATUS OF IMPLEMENTATION (CONTINUED)
STANDBYLIQUIDCONTROLSYSiEMMODIFICATIONS
~
INSTALLED AT 6 PLANTS (NT0Ls) 3 SCHEDULED IN 1986 13 SCHEDULED IN 1987 9 SCHEDULED IN 1988 LACROSSE, VERMONT YANKEE TO REQUEST EXEMPTION
\\
BROWNS FERRY, SHOREHAM, FERMI-2 SCHEDULES UNKNOWN DUE TO STARTUP DATE UNCERTAINTIES l
l l
l l
i l
26
r-go REVIEW 0F EPGs FOR ATWS REVISION 2 APPROVED FEB 1983 NO SIGNIFICANT CHANGES ~FOR ATWS PROCEDURES IN REV. 4 FOR EXTREME CONDITIONS EPGs DIRECT OPERATOR T0:
INHIBIT ADS LOWER WATER LEVEL TO TAF TO REDUCE POWER WHILE BORON IS INJECTED RAISE WATER LEVEL AFTER BORON IS INJECTED TO ENHANCE BORON MIXING l
l I
9